• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.7
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• pp.843-850
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• 2010

A spacer grid assembly is one of the most important structural components of the nuclear fuel assembly of a Pressurized Water Reactor (PWR). A primary design requirement is that the fuel rod integrity be maintained by the spacer grid assembly during the operation of the reactor. In this study, we suggested a new spacer grid assembly having a fuel rod support, which is capable of sliding when the fuel rod vibrates due to flow-induced vibrations in the reactor. By adjusting the relative displacement between the fuel rod and its support, the proposed design will help in reducing fuel rod fretting damage.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.35 no.2
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• pp.201-206
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• 2011

A spacer grid assembly is one of the most important structural components of the nuclear fuel assembly of a pressurized water reactor (PWR), and it affects the performance of the fuel assembly. The primary design requirement is that the mechanical integrity of the fuel rod should be maintained by the spacer grid assembly during the operation of the reactor. It was known that fretting damage to the fuel rod can be reduced by adjusting the relative moving displacement between the fuel rod and its support. In this study, we used the finite element method to evaluate the characteristics of a sliding support designed to reduce fretting damage of fuel rods.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.36 no.7
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• pp.689-695
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• 2012

The friction factor in a rod bundle and the loss coefficient at a spacer grid were examined. As a test section, 25 smooth rods, 9.5 mm in diameter and 2000 mm in length, were prepared and installed in a $5{\times}5$ square array in a square channel. In this case, the P/D (Pitch-to-Diameter ratio) was 1.35. In this work, plain (i.e., no mixing vanes), split-vane, and hybrid-vane spacer grids were tested. In a bare rod bundle (i.e., no spacer grid), the measured friction factors were in good agreement with the previous correlations. Among the spacer grids tested, the hybrid-vane spacer grid presented the largest friction factor in the rod bundle and loss coefficient. This may be because of the flow pattern change induced by large relative plugging of the flow cross section and mixing vane geometry. At Re=$5{\times}10^5$, the predicted loss coefficients of plain, splitvane, and hybrid-vane spacer grids were approximately 0.79, 0.80, and 0.88, respectively.

• Nuclear Engineering and Technology
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• v.28 no.1
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• pp.17-26
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• 1996

The in-reactor fuel rod support conditions against the fretting wear-induced damage can be evaluated by residual spacer grid spring deflection or rod-to-grid gap. In order to evaluate the impact of fuel design parameters on the fretting wear-induced damage, a simulation methodology of the in-reactor fuel rod supporting conditions as a function of burnup has been developed and implemented in the GRIDFORCE program. The simulation methodology takes into account cladding creep rate, initial spring deflection, initial spring force, and spring force relaxation rate as the key fuel design parameters affecting the in-reactor fuel rod supporting conditions. Based on the parametric studies on these key parameters, it is found that the initial spring deflection, the spring force relaxation rate and cladding creepdown rate are in the order of the impact on the in-reactor fuel rod supporting conditions. Application of this simulation methodology to the fretting wear-induced failure experienced in a commercial plant indicates that this methodology can be utilized as an effective tool in evaluating the capability of newly developed cladding materials and/or new spacer grid designs against the fretting wear-induced damage.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.297-302
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• 1998

유한요소법을 이용하여 스프링으로 연속 지지되고 축방향 하중이 작용하는 핵연료봉의 자유진동 해석올 수행하였다. 본 해석에는 지지격자 지지점에서 핵연료봉의 변위가 구속되지 않는 실제 경계조건을 반영하였다. 이러한 경계조건은 지지점 스프링 상수에 의하여 핵연료봉 해석모델의 탄성항이 약화되는 현상을 반영할 수 있어서 지지점이 구속된 기존의 모델보다 고유진동수를 작게 예측한다. 스프링 상수가 어떤 임계값 이하를 갖는 경우 고유진동수 뿐만 아니라 모드형상도 크게 변하기 때문에 지지점을 구속한 모델에 의한 해석은 실제 진동현상을 상당히 왜곡 할 수 있다. 핵연료봉에 작용하는 축방향력이 인장력에서 압축력으로 감소함에 따라 고유진동수도 감소하지만 핵연료봉의 고유형상은 변하지 않았다. 지지격자 스프링 상수의 점진적인 감소와 핵연료봉 축방향 압축력의 감소를 동시에 적용하는 경우 고유 진동수는 두 변수를 별도로 적용했을 때 얻은 최소값의 변화에 따르는 경향을 나타내었다.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.3
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• pp.157-164
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• 2016

The performance of convective heat transfer in rod bundle flow was experimentally evaluated using a twist-vane spacer grid. A $4{\times}4$ square-arrayed rod bundle was prepared as the test section, with a pitch-to-diameter ratio(P/D) of ~1.35. To check the convective heat transfer performance, the circumferential and longitudinal variations in rod-wall temperatures were measured downstream of the twist-vane spacer grid. In the circumferential measurements, the rod-wall temperature toward the twist-vane tip showed the lowest value, which might be due to the deflected water flow caused by the twist-vane. On the other hand, the wall temperature of the longitudinal measurements near the twist-vane spacer grid decreased dramatically, which implies that the convective heat transfer performance was enhanced. A heat transfer enhancement of ~35 % was achieved near downstream of the twist-vane spacer grid, as compared with the upstream value. Based on the present experimental data, a correlation for predicting the heat transfer performance of a twist-vane spacer grid was proposed.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05b
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• pp.291-296
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• 1998

노내에서 지지격자 스프링의 잔류 탄성변위는 시간(연소도)에 따라 변하게 된다. 이는 격자판의 중성자 조사에 의한 길이방향의 성장으로 지지격자 셀 크기의 증가와 피복관의 크리프에 의한 직경의 감소 및 중성자 조사에 의한 지지격자 스프링력의 이완으로 인한 것이다. 만일 지지격자 스프링의 거동이 변하여 연료봉을 탄성적으로 지지하지 못할 경우 이것은 연료봉의 유체에 의한 진동을 가속시키게 되며, 연료봉과 지지격자 스프링이나 딤플간의 반복적인 고주기의 충격하중은 연료봉의 지지부와 봉간(grid-to-rod)의 프레팅 마모의 원인이 될 수 있다. 따라서 시간에 따라 변하는 변수들의 영향을 고려한 지지격자 스프링의 잔류 탄성변위를 예측할 수 있는 방법론을 정립하여 새로운 지지격자체의 개발시 건전한 연료봉의 지지거동을 평가할 수 있는 도구로 활용하고자 하였다.

• Nuclear Engineering and Technology
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• v.28 no.2
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• pp.162-174
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• 1996

The local hydraulic characteristics in a single phase flow of a 6$\times$6 rod bundle with neighboring different spacer grids were measured by using a LDV(Laser Doppler Velocimeter) system. 6$\times$6 rod bundle is formed by two 3$\times$6 rod bundles with different spacer grids. The objective of this study in a rod bundle is to investigate the thermal-hydraulic interactions between different spacer grids with different configurations and resistance. By using a LDV system, the velocity and turbulent intensity in axial and horizontal directions ore measured. Pressure drop measurements ore also performed to evaluate the loss coefficient for the spacer grid and the friction factor for rod bundles. Implications concerning thermal mining due to spacer grids were investigated based on the hydraulic test results. Swirl factor, which is assumed as a qualitative criteria for DNB(departure from nucleate boiling), was defined and estimated from the horizontal velocity result.

• Proceedings of the Korean Society for Noise and Vibration Engineering Conference
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• 2007.05a
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• pp.361-365
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• 2007

우라늄을 내장한 연료봉은 핵분열이 일어나는 우라늄 펠렛(pellet)을 1차적으로 차폐하는 중요한 구조물이다. 연료봉은 원자로 내에서 유체유발진동에 의해 손상될 수 있으며, 본 연구에서는 유동유발진동 특성을 예측하기 위해 핵연료봉의 동특성 규명을 위한 모드해석을 수행하였다. 핵연료봉의 진동특성을 규명하기 위해 제작한 시험장치를 이용하여 피복관(clad tube)의 진동특성실험과 유한 요소 해석을 수행하였다. 모드시험(Modal Testing)은 현재 상용 핵연료봉(튜브)을 대상으로 수행되었으며, 유한 요소 해석 모델을 개발하여 해석 결과와 시험 결과를 비교 분석하였다.

• Nuclear Engineering and Technology
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• v.28 no.1
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• pp.56-71
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• 1996

The local hydrulic characteristics in subchannels of 5$\times$5 nuclear fuel bundles with spacer grids were measured at upstream and downstream of the spacer grid for the investigation of the spacer grid effects on turbulent flow structure by using an LDV(Laser Doppler Velocimeter). The measured parameters are axial velocity and turbulent intensity, skewness factor, and flatness factor. Pressure drops were also measured to evaluate the loss coefficient for the spacer grid and the friction factor for rod bundles. From these data, it was found that the turbulent mixing and forced mixing occur up to $x/D^h=10$ and 20 from the spacer grid, respectively. The turbulence decay behind spacer grid behaves in the similar decay rate as turbulent flow through mesh grids or screens. Mixing factors useful in subchannel analysis code were correlated from the data and show the highest value near spacer grid and then have a stable values.